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OLIVEIRA, JOSE R. de. "Programa computacional para estudo da estrategia de controle de um reator nuclear do tipo PWR". reponame:Repositório Institucional do IPEN, 2002. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11060.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Huntington, James E. "Computer simulation studies of nuclear reactor fuel and related uranium phases". Thesis, Keele University, 1994. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.384936.
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Biaty, Patricia Andrea Paladino. "Pré-processador matemático para o código Relap5 utilizando o Microsoft Excel". Universidade de São Paulo, 2006. http://www.teses.usp.br/teses/disponiveis/85/85133/tde-14052007-141446/.
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O estudo termo-hidráulico, utilizado para análise de acidentes e transientes em reatores nucleares, é feito com o uso de algumas ferramentas computacionais sofisticadas. Esses programas utilizam uma filosofia realista (best estimate) para análise de acidentes e transientes em reatores refrigerados à água leve do tipo PWR (Pressurized Water Reactor) e sistemas associados. O código RELAP5, objeto de nosso estudo, tem sido usado como uma ferramenta para o licenciamento de instalações nucleares no nosso país. Uma das maiores dificuldades na simulação de acidentes e transientes em uma instalação nuclear com o código RELAP5 é a quantidade de informações necessárias, que na maioria dos casos é muito grande. Além disso, existe a necessidade de uma quantidade razoável de operações matemáticas para os cálculos da geometria dos componentes. Portanto, a fim de facilitar a manipulação destas informações, percebeu-se a necessidade do desenvolvimento de um pré-processador amigável com o usuário, para realização desses cálculos e para elaboração dos dados de entrada do RELAP5. A ferramenta escolhida foi o MS-EXCEL, que apresentou grande potencialidade no desenvolvimento do pré-processador desejado.Computational program are used for thermal hydraulic analysis of accidents and transients conditions in nuclear power plants. The RELAP5 code has been developed to simulate accidents and transients conditions, performing a best estimate analysis, in Pressurized Water Reactors (PWR) and auxiliary systems. The RELAP5 code, which has been used as a tool for licensing nuclear facilities in Brazil, is the objective of the study performed in this work. The main problem in using the RELAP5 code is the huge amount of information necessary to model the nuclear reactor and thus to simulate thermal-hydraulic accidents. Moreover, the RELAP5 code input data requires a large amount of mathematical operations to calculate the geometry of the plant components. Therefore, in order to make easier the data input for the RELAP5 code a friendly preprocessor has been developed. The preprocessor accepts basic information about the geometry of the plant components and performs all the calculations needed for the RELAP5 input. This preprocessor has been developed based on the MS-EXCEL software.
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Dalle, Hugo Moura. "Simulação do reator TRIGA IPR-R1 utilizando metodos de transporte por Monte Carlo". [s.n.], 2005. http://repositorio.unicamp.br/jspui/handle/REPOSIP/267210.
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Orientador: Elias Basile TambourgiTese (doutorado) - Universidade Estadual de Campinas, Faculdade de Engenharia Quimica
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Resumo: A utilização do método Monte Carlo na simulação do transporte de partículas em reatores nucleares é crescente e constitui uma tendência mundial. O maior inconveniente dessa técnica, a grande exigência de capacidade de processamento, vem sendo superado pelo contínuo desenvolvimento de processadores cada vez mais rápidos. Esse contexto permitiu o desenvolvimento de metodologias de cálculo neutrônico de reatores nas quais se acopla a parte do transporte de partículas, feita com um código de Monte Carlo, ao cálculo de queima e decaimento radioativo. Neste trabalho tal metodologia de simulação é implantada, validada para reatores de pesquisas, notadamente os do tipo TRIGA e finalmente utilizada na simulação neutrônica do reator TRIGA IPR ¿ RI do CDTN/CNEN. O sistema de códigos empregados é constituído pelos amplamente utilizados códigos MCNP4B (transporte por método Monte Carlo) e ORIGEN2.1 (queima e decaimento radioativo). Apesar dos esforços recentes no sentido de agrupar as duas etapas de cálculo, transporte e queima, em um único código, até o momento esta opção não está disponível e, portanto, um terceiro código é utilizado para realizar o acoplamento transporte/queima. Neste trabalho utilizou-se para tal o código MONTEBURNS. O sistema formado por estes três códigos permitiu obter os parâmetros neutrônicos de interesse do IPR ¿ R1 através apenas de simulação teórica, sem a necessidade de qualquer tipo de ajuste baseado em dados experimentais, em boa concordância com os valores medidos... Observação: O resumo, na íntegra, poderá ser visualizado no texto completo da tese digital
Abstract: The use of Monte Carlo methods in particles transport simulations of nuclear reactor is growing fast and constitutes a strong tendency all over the world. The major inconvenient of such techniques is the huge demand of processing power which has been surpassed the development of reactor physics calculation methodologies in which the particles transport part, made by a Monte Carlo transport code, is linked with the burnup and radioactive decay part of the simulation. On this work a such simulation methodology is made operational, validated for research reactors, mainly for TRIGA reactor and finally utilized for reactor physics simulation of the CDTN¿s TRIGA IPR ¿ R1. The adopted codes system is constituted by the widespreadly used codes MCNP4B (Monte Carlo transport) and ORIGEN2.1 (burnup and radioactive decay). In spite of the very recent efforts toward get together both, transport and burnup, in only one code at the moment this is a not available option and therefore, a third code is needed to carry out the linkage transport/burnup. MONTEBURS code was used to this purpose. This three codes system has allowed to obtain the physical parameters of IPR ¿ R1 calculated using only theoretical simulation without any kind of experimental adjustment or interaction between experiments and calculation in good agreement with measured values... Note: The complete abstract is available with the full electronic digital thesis or dissertations
Doutorado
Sistemas de Processos Quimicos e Informatica
Doutor em Engenharia Química
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Bollen, Rob. "Verification and validation of computer simulations with the purpose of licensing a pebble bed modular reactor". Thesis, Stellenbosch : Stellenbosch University, 2002. http://hdl.handle.net/10019.1/53216.
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Thesis (MBA)--Stellenbosch University, 2002.ENGLISH ABSTRACT: The Pebble Bed Modular Reactor is a new and inherently safe concept for a nuclear power generation plant. In order to obtain the necessary licenses to build and operate this reactor, numerous design and safety analyses need to be performed. The results of these analyses must be supported with substantial proof to provide the nuclear authorities with a sufficient level of confidence in these results to be able to supply the required licences. Beside the obvious need for a sufficient level of confidence in the safety analyses, the analyses concerned with investment protection also need to be reliable from the investors’ point of view. The process to be followed to provide confidence in these analyses is the verification and validation process. It is aimed at presenting reliable material against which to compare the results from the simulations. This material for comparison will consist of a combination of results from experimental data, extracts from actual plant data, analytical solutions and independently developed solutions for the simulation of the event to be analysed. Besides comparison with these alternative sources of information, confidence in the results will also be built by providing validated statements on the accuracy of the results and the boundary conditions with which the simulations need to comply. Numerous standards exist that address the verification and validation of computer software, for instance by organisations such as the American Society of Mechanical Engineers (ASME) and the Institute of Electrical and Electronics Engineers (IEEE). The focal points of the verification and validation of the design and safety analyses performed on typical PBMR modes and states, and the requirements imposed by both the local and overseas nuclear regulators, are not entirely enveloped by these standards. For this reason, PBMR developed a systematic and disciplined approach for the preparation of the Verification and Validation Plan, aimed at capturing the essence of the analyses. This approach aims to make a definite division between software development and the development of technical analyses, while still using similar processes for the verification and validation. The reasoning behind this is that technical analyses are performed by engineers and scientists who should only be responsible for the verification and validation of the models and data they use, but not for the software they are dependent on. Software engineers should be concerned with the delivery of qualified software to be used in the technical analyses. The PBMR verification and validation process is applicable to both hand calculations and computer-aided analyses, addressing specific requirements in clearly defined stages of the software and Technical Analysis life cycle. The verification and validation effort of the Technical Analysis activity is divided into the verification and validation of models and data, the review of calculational tasks, and the verification and validation of software, with the applicable information to be validated, captured in registers or databases. The resulting processes are as simple as possible, concise and practical. Effective use of resources is ensured and internationally accepted standards have been incorporated, aiding in faith in the process by all stakeholders, including investors, nuclear regulators and the public.
AFRIKAASE OPSOMMING: Die Modulêre Korrelbedreaktor is ’n nuwe konsep vir ’n kernkragsentrale wat inherent veilig is. Dit word deur PBMR (Edms.) Bpk. ontwikkel. Om die nodige vergunnings om so ’n reaktor te kan bou en bedryf, te bekom, moet ’n aansienlike hoeveelheid ontwerp- en veiligheidsondersoeke gedoen word. Die resultate wat hierdie ondersoeke oplewer, moet deur onweerlegbare bewyse ondersteun word om vir die owerhede ’n voldoende vlak van vertroue in die resultate te gee, sodat hulle die nodigde vergunnings kan maak. Benewens die ooglopende noodsaak om ’n voldoende vlak van vertroue in die resultate van die veiligheidsondersoeke te hê, moet die ondersoeke wat met die beskerming van die beleggers se beleggings gepaard gaan, net so betroubaar wees. Die proses wat gevolg word om vertroue in die resultate van die ondersoeke op te bou, is die proses van verifikasie en validasie. Dié proses is daarop gerig om betroubare vergelykingsmateriaal vir simulasies voor te lê. Hierdie vergelykingsmateriaal vir die gebeurtenis wat ondersoek word, sal bestaan uit enige kombinasie van inligting wat in toetsopstellings bekom is, wat in bestaande installasies gemeet is, wat analities bereken is; asook dit wat deur ’n derde party onafhanklik van die oorspronklike ontwikkelaars bekom is. Vertroue in die resultate van die ondersoeke sal, behalwe deur vergelyking met hierdie alternatiewe bronne van inligting, ook opgebou word deur die resultate te voorsien van ’n gevalideerde verklaring wat die akkuraatheid van die resultate aantoon en wat die grensvoorwaardes waaraan die simulasies ook moet voldoen, opsom. Daar bestaan ’n aansienlike hoeveelheid internasionaal aanvaarde standaarde wat die verifikasie en validasie van rekenaarsagteware aanspreek. Die standaarde kom van instansies soos die Amerikaanse Vereniging vir Meganiese Ingenieurs (ASME) en die Instituut vir Elektriese en Elektroniese Ingenieurs (IEEE) – ook van Amerika. Die aandag wat deur die Suid-Afrikaanse en oorsese kernkragreguleerders vereis word vir die toestande wat spesifiek geld vir korrelbedreaktors, word egter nie geheel en al deur daardie standaarde aangespreek nie. Daarom het die PBMR maatskappy ’n stelselmatige benadering ontwikkel om verifikasie- en validasieplanne voor te berei wat die essensie van die ondersoeke kan ondervang. Hierdie benadering is daarop gemik om ’n duidelike onderskeid te maak tussen die ontwikkeling van sagteware en die ontwikkeling van tegniese ondersoeke, terwyl steeds gelyksoortige prosesse in die verifikasie en validasie gebruik sal word. Die rede hiervoor is dat tegniese ondersoeke uitgevoer word deur ingenieurs en wetenskaplikes wat net vir verifikasie en validasie van hulle eie modelle en die gegewens verantwoordelik gehou kan word, maar nie vir die verifikasie en validasie van die sagteware wat hulle gebruik nie. Ingenieurs wat spesialiseer in sagteware-ontwikkeling behoort verantwoordelik te wees vir die daarstelling van sagteware wat deur die reguleerders gekwalifiseer kan word, sodat dit in tegniese ondersoeke op veiligheidsgebied gebruik kan word. Die verifikasie- en validasieproses van die PBMR is sowel vir handberekeninge as vir rekenaarondersteunde-ondersoek geskik. Hierdie proses spreek spesifieke vereistes in onderskeie stadiums gedurende die lewenssiklusse van die ontwikkeling van sagteware en van tegniese ondersoeke aan. Die verifikasie- en validasiewerk vir tegniese ondersoeksaktiwiteite is verdeel in die verifikasie en validasie van modelle en gegewens, die nasien van berekeninge en die verifikasie en validasie van sagteware, waarby die betrokke inligting wat gevalideer moet word, versamel word in registers of databasisse. Die prosesse wat hieruit voortgevloei het, is so eenvoudig as moontlik, beknop en prakties gehou. Hierdeur is ’n effektiewe benutting van bronne verseker. Internasionaal aanvaarde standaarde is gebruik wat die vertroue in die proses deur alle betrokkenes, insluitende beleggers, die owerhede en die publiek, sal bevorder.
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Revel, Aldric. "Nuclear forces at the extremes". Thesis, Normandie, 2018. http://www.theses.fr/2018NORMC227/document.
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L’émission de paires de neutrons par les noyaux riches en neutrons 18C et 20O (isotones N = 12) est étudié par réactions de knock-out d’un nucléon des faisceaux secondaires 19N et 21O, peuplant ainsi des états non liés jusqu’à 15 MeV au-dessus de leur seuil d’émission deux neutrons. L’analyse des corrélations des triples coïncidences fragment+n+n montre que la décroissance 19N(−1p)→18C → 16C+n+n est clairement dominée par l’émission directe de paires. Les corrélations n-n, les plus grandes jamais observées, suggèrent la prédominance d’un coeur de 14C entouré de quatre neutrons arrangés en paires très corrélées. De plus, une importante compétition du mode de décroissance séquentiel est observée dans la décroissance 21O(−1n) → 20O → 18O+n+n, interprétée par la déformation causée par le knock-out d’un neutron très lié ayant pour effet de casser le cœur de 16O et ainsi de réduire le nombre de paires.De plus, les états non liés du 26F et 28F sont étudiés. Les deux systèmes étant peuplés par knock-out d’un nucléon du 27F dans le cas du 26F et du 29Ne ou du 29F pour 28F. Cinq états ont été observés pour 26F avec en particulier l’état de plus basse énergie (0.39 MeV) identifié comme l’état 3+ résultant du couplage d5/2 ⊗ d3/2 . Pour 28F, cinq états ont aussi été observés et l’état fondamental (200 keV) a été identifié comme étant de parité négative, plaçant ainsi 28F dans l’îlot d’inversionThe emission of neutron pairs from the neutron-rich N = 12 isotones 18C and 20O has been studied by high-energy nucleon knockout from 19N and 21O secondary beams, populating unbound states of the two isotones up to 15 MeV above their two-neutron emission thresholds. The analysis of triple fragment-n-n correlations shows that the decay 19N(−1p) → 18C → 16C+n+n is clearly dominated by direct pair emission. The two-neutron correlation strength, the largest ever observed, suggests the predominance of a 14C core surrounded by four neutrons arranged in strongly correlated pairs. On the other hand, a significant competition of a sequential branch is found in the decay 21O(−1n) → 20O → 18O+n+n, attributed to its formation through the knockout of a deeply-bound neutron that breaks the 16O core and reduces the number of pairs.Moreover, unbound states in 26F and 28F have been studied. The two systems were probed using single-nucleon knockout reaction from secondary beams of 27F respectively in the case of 26F, and 29Ne and 29F for 28F. Five possible states have been identified in 26F, with in particular the lowest energy one (0.39 MeV) being identified as the 3+ state resulting from the d5/2 ⊗ d3/2 coupling. In the case of 28F, five unbound state have also been observed and in particular its ground state (200 keV) has been identified as a negative parity state, meaning that 28F is located inside the island of inversion
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Tshamala, Mubenga Carl. "Simulation and control implications of a high-temperature modular reactor (HTMR) cogeneration plant". Thesis, Stellenbosch : Stellenbosch University, 2014. http://hdl.handle.net/10019.1/86264.
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Thesis (MScEng)--Stellenbosch University, 2014.ENGLISH ABSTRACT: Traditionally nuclear reactor power plants have been optimised for electrical power generation only. In the light of the ever-rising cost of dwindling fossil fuel resources as well the global polluting effects and consequences of their usage, the use of nuclear energy for process heating is becoming increasingly attractive. In this study the use of a so-called cogeneration plant in which a nuclear reactor energy source is optimised for the simultaneous production of superheated steam for electrical power generation and process heat is considered and analysed. The process heat superheated steam is generated in a once-through steam generator of heat pipe heat exchanger with intermediate fluid while steam for power generation is generated separately in a once-through helical coil steam generator. A 750 °C, 7 MPa helium cooled HTMR has been conceptually designed to simultaneously provide steam at 540 °C, 13.5 MPa for the power unit and steam at 430 °C, 4 MPa for a coal-to-liquid fuel process. The simulation and dynamic control of such a typical cogeneration plant is considered. In particular, a theoretical model of a typical plant will be simulated with the aim of predicting the transient and dynamic behaviour of the HTMR in order to provide guideline for the control of the plant under various operating conditions. It was found that the simulation model captured the behaviour of the plant reasonably well and it is recommended that it could be used in the detailed design of plant control strategies. It was also found that using a 1500 MW-thermal HTMR the South African contribution to global pollution can be reduced by 1.58%.
AFRIKAANSE OPSOMMING: Tradisioneel is kernkragaanlegte vir slegs elektriese kragopwekking geoptimeer. In die lig van die immer stygende koste van uitputbare fossielbrandstohulpbronne asook die besoedelingsimpak daarvan wêreldwyd, word die gebruik van kernkrag vir prosesverhitting al hoe meer aanlokliker. In hierdie studie word die gebruik van ‘n sogenaamde mede-opwekkingsaanleg waarin ‘n kernkragreaktor-energiebron vir die gelyktydige produksie van oorverhitte stoom vir elektriese kragopwekking en proseshitte oorweeg ontleed word. Die oorvehitte stoom word in ‘n enkeldeurvloei-stoomopwekking van die hittepyp-hitteruiler met tussenvloeistof opgewek en stoom vir kragopwekking word apart in ‘n enkeldeurvloei-spiraalspoel-stoomopwekker opgewek. ‘n 750 °C, 7 MPa heliumverkoelde HTMR is konseptueel ontwerp vir die gelytydige veskaffing van stoom by 540 °C, 13.5 MPa, vir die kragopwekkings eenheid, en stoom by 430 °C, 4 MPa, vir ‘n steenkool-tot-vloeibare (CTL) brandstoff proses. Die simulasie en dinamiese beheer van ‘n tipiese HTMR mede-opwekkingsaanleg word beskou. ‘n die besonder word ‘n teoretiese model van die transiënte en dinamiese gedrag van die aanleg gesimuleer om sodoene riglyne te identifiseer vir die ontwikkeling van dinamiese beheer strategië vir verskillende werkstoestande van die aanleg. Daar was ook gevind dat die simulasie model van die aanleg se gedrag goed nageboots word en dat dit dus gebruik kan word vir beheer strategie doeleindes. Indien so ‘n 1500 MW-termies HTMR gebruik word sal dit die Suid Afrikaanse besoedling met 1.58% sal kan verminder.
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SILVESTRE, LARISSA J. B. "PCRELAP5 - Programa de cálculo para os dados de entrada do código RELAP5". reponame:Repositório Institucional do IPEN, 2016. http://repositorio.ipen.br:8080/xmlui/handle/123456789/26393.
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Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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PERRENOUD, HELENA G. "Modulo de extracao de eventos em assinaturas de potencia de valvulas moto-operadas, usando um sistema especialista para o sistema de diagnostico de MOV's utilizado em reatores nucleares". reponame:Repositório Institucional do IPEN, 2001. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10967.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Kruckenberg, Norman E. "A piping network model program for small computer". Ohio : Ohio University, 1986. http://www.ohiolink.edu/etd/view.cgi?ohiou1183138191.
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Middleburgh, Simon C. "Atomistic scale simulation of materials for future nuclear reactors". Thesis, Imperial College London, 2012. http://hdl.handle.net/10044/1/9615.
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Atomic scale simulations have been carried out on three systems that are being considered for use in future nuclear energy applications, both fission and fusion based. Uranium dioxide and chromium doped fuel are considered in the early chapters in order to understand the processes important in high burnup nuclear fuel. The oxygen stoichiometry of the uranium dioxide lattice was found to have a large effect on both fission product solution and crystal swelling. Predictions were found to replicate experimental data well. Transport properties of cations via uranium vacancies in hyperstoichiometic UO2+x have been studied for the first time on the atomic scale. Understanding the arrangement of U5+ cations around a migrating species has proved important for identifying low energy migration process. Zirconium diboride and beryllium have also been studied. Zirconium diboride is of interest due to its use as a burnable poison for some advanced fuel types and also because of its ability to resist very high temperatures. The variation in stoichiometry of ZrB2 was found to accommodate excess boron but very little excess zirconium. The accommodation of the boron-10 transmutation products, lithium and helium, are also studied with helium being released from the lattice via a low energy process. Beryllium is of importance as a potential cladding for fission fuel and in fusion reactors. The intrinsic defect behaviour has been discussed for the first time in this thesis while extrinsic species present in beryllium alloys through alloying, manufacturing processes or environmental exposure have also been studied. Again, helium was found to be readily released from the lattice but only as an interstitial species and not as a substitutional defect.
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Khamis, Ibrahim Ahmad. "Simulation of nuclear power plant pressurizers with application to an inherently safe reactor". Diss., The University of Arizona, 1988. http://hdl.handle.net/10150/184378.
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Pressurizer modeling for predicting the dynamic pressure of the PIUS system is presented. The transient behavior of this model for the PIUS system was investigated. The validity of this model for the PIUS system is limited to transients that are neither too large nor too long in duration. For example, the model is not capable of describing events following a complete loss of liquid for the pressurizer. However, the model can be used for qualitative prediction of the PIUS system behavior for a wide variety of severe transients. A review of pressurizer modeling indicates that the neglecting of the change in the internal energy of the subcooled water during transients is an acceptable assumption. The inherently safe feature of the PIUS system was confirmed through the self-shutdown of the reactor or, in some cases, through reactor power reduction as a result of the ingress of the pool boric acid solution into the primary system. This dynamic model was constructed of three major components: (1) The primary loop, (2) The secondary loop, and (3) The natural convection loop through the pool. A lumped parameter model, uniform heat transfer, and point kinetics have been the main approximations in this model. Other approximations are mentioned during the modeling of each component of the model. The dynamic model was simulated using the DARE-P continuous system simulation language which was developed in the Electrical Engineering Department at the University of Arizona.
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Monteagudo, Godoy Belen. "Structure and neutron decay of the unbound Beryllium isotopes 15,16Be". Thesis, Normandie, 2019. http://www.theses.fr/2019NORMC254.
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A la limite de stabilité nucléaire, où l'énergie nécessaire pour enlever un nucléon tend vers zéro, l'émission de neutrons est un phénomène caractéristique des noyaux riches en neutrons. La désintégration "deux neutrons" d'un noyau à l'état fondamental est un cas particulier qui peut se produire au-delà de cette limite. La spectroscopie et la désintégration des isotopes les plus riches en neutrons du béryllium 14Be, 15Be et surtout 16Be ont été étudiées lors de différentes campagnes expérimentales au RIBF-RIKEN en utilisant le dispositif SAMURAI couplé au détecteur de neutrons NEBULA. Dans ce dernier cas, la cible MINOS a été ajoutée à la configuration standard.L'approche expérimentale et un traitement approfondi des événements multi-neutron ont rendu possible une analyse détaillée des corrélations à trois corps fragment+n+n à partir d'états non liés à deux neutrons. Pour la première fois, l’état fondamental et le premier état excité du 16Be ont été observés sans ambiguïté, et les analyses des corrélations montrent clairement une émission directe de paires de neutrons a partir des deux états. Afin d'interpréter les distributions expérimentales, l'émission de paires de neutrons (nn) a été caractérisée à partir d'une approche théorique microscopique. La comparaison directe entre les résultats expérimentaux et les prédictions théoriques a permis de relier le signal expérimental nn que nous observons à la fonction d'onde à trois corpsNear the dripline, where the energy needed to remove one nucleon is low enough or even negative, neutron emission from neutron-rich nuclei is a characteristic phenomenon. Ground state two-neutron decays are a special case that may occur once we go beyond the dripline. The spectroscopy and neutron decay of the most neutron-rich isotopes of Beryllium 14Be, 15Be and especially 16Be have been investigated during different experimental campaigns at RIBF-RIKEN using the SAMURAI setup and the NEBULA neutron array. For the latter, the state-of-the-art MINOS target was added to the standard setup.Our experimental approach and a thorough treatment of multi-neutron events have made possible an extensive analysis on the three-body correlations in fragment+n+n decays from two-neutron unbound states. For the first time, the ground state and first excited state of 16Be have been unambiguously observed, and the correlation analyses show a clear direct neutron-pair emission from both states. In order to interpret the experimental distributions, the neutron-pair emission (nn) has been characterized from microscopic principles. A joint effort between experiment and theory has allowed a direct comparison to connect the experimental nn signal we observe with the three-body wave-function
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Cue, R. J. "Multiprocessing in nuclear plant simulation". Thesis, University of Manchester, 1987. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.382772.
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Sadeghi, Mohammad Mehdi 1959. "SYMBOLIC MANIPULATION IN REACTOR PHYSICS". Thesis, The University of Arizona, 1986. http://hdl.handle.net/10150/275520.
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Enani, Mohammad A. "MURR nodal analysis with simple interactive simulation /". free to MU campus, to others for purchase, 1997. http://wwwlib.umi.com/cr/mo/fullcit?p9841283.
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Adams, Joseph T. "Neural network calibration of moderator temperature coefficient measurements in pressurized water nuclear reactors". Thesis, This resource online, 1993. http://scholar.lib.vt.edu/theses/available/etd-12042009-020108/.
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Trosman, Hernan Gerardo. "Computer simulation for transient analysis of MITR loop components". Thesis, Massachusetts Institute of Technology, 1994. http://hdl.handle.net/1721.1/12130.
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Sumner, Tyler. "Effects of fuel type on the safety characteristics of a sodium cooled fast reactor". Diss., Georgia Institute of Technology, 2010. http://hdl.handle.net/1853/37217.
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A series of accident simulations were performed using INL's thermal hydraulics code RELAP5-3D to analyze steady-state and transient behavior of a sodium cooled fast reactor. The reactor chosen for this study was General Electric's S-PRISM, which is a 1,000 MWt pool-type sodium-cooled fast reactor, designed for either an Oxide or Metal fueled core. Once key core characteristics including power profiles, reactivity feedback coefficients and delayed neutron parameters were calculated, S-PRISM was redesigned for a Nitride fueled core to take advantage of the Nitride fuel's high thermal conductivity and melting temperature. Loss of flow, loss of heat sink, loss of power and inadvertent control rod withdrawal accidents were simulated for each core at beginning, middle and end of cycle to determine if one fuel type provides significant safety advantages over the others.
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Zhang, Yi 1979. "Computer simulation and topological modeling of radiation effects in zircon". Thesis, Massachusetts Institute of Technology, 2006. http://hdl.handle.net/1721.1/41587.
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Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006.Includes bibliographical references.
The purpose of this study is to understand on atomic level the structural response of zircon (ZrSiO4) to irradiation using molecular dynamics (MD) computer simulations, and to develop topological models that can describe these structural changes. Topological signatures, encoded using the concepts of primitive-rings and local clusters, were developed and used to differentiate crystalline and non-crystalline atoms in various zircon structures. Since primitive-rings and local clusters are general concepts applicable to all materials, and the algorithms to systematically identify them are well-established, topological signatures based on them are easy to implement and the method of topological signatures is applicable to all structures. The method of topological signatures is better than the Wigner-Seitz cell method, which depends on the original crystalline reference grid that is unusable in heavily damaged structures or regions; it is also better than those methods based only on local structures limited to first coordination shell, since one can decide whether or not to include ring contents of large rings into the topological signatures, effectively controlling the range of the topological signatures. The early-stage evolution of non-crystalline disorder and the subsequent recrystallization in zircon collision cascade simulations were successfully modeled by using the topological signatures to identify non-crystalline atoms. Simply using the number of displaced atoms was unable to correctly show the initial peak of structural damage followed by the subsequent annealing stage. Using the topological signatures, amorphization within a single collision cascade was observed in zircon.
(cont.) In the radiation-induced amorphous zircon simulated in this study, the method of topological signatures was able to differentiate the amorphous region in the center of the simulation box and the crystalline region surrounding it. A few isolated remnant crystalline islands were identified in the amorphous region. About 5% of atoms in melted and melt-quenched structures were identified as crystalline atoms. Different amorphous zircon structures were found to be topologically different. Upon amorphization of zircon, the average ring size and the number of atoms in local cluster were found to increase. Larger average ring sizes were found in more pervasively amorphized structures. The radiation-induced amorphous structure was the least pervasively amorphized one, followed by the melt-quenched. The liquid-state amorphous structure was most pervasively amorphized and had the largest average ring size. Phase-separation of zircon into SiO2- and ZrO2-rich local regions was observed when zircon was amorphized in simulations, either thermally or by radiation. It was found in simulations using constant pressure ensembles that the zircon structure underwent abnormally huge volume swelling when it amorphized, which was attributed to the ion charges used in the potential model. Although the ion charges used in the originally chosen potential model were overall balanced, they were not balanced with regard to the phase decomposition products, and thus resulted in strong Coulombic repulsive force within locally SiO2- and ZrO2-rich regions when phase separation occurred. After the ion charges were re-balanced (and other potential parameters refitted), the volume expansion was found to be under control. The charge imbalance of SiO2 units was also found to produce unrealistically large fraction of 3-coordinated Si and shorter Si-O bond length.
(cont.) The issue of charge-balance with regard to phase decomposition products applies to all complex ceramics that decompose into separate phases upon amorphization. Threshold displacement energies in zircon were systematically determined. Many special directions, such as those directed toward neighboring atoms or open spaces surrounding the PKA, were considered. Cascade detail was extensively examined, including PKA trajectory, cascade extent, time scale, thermal spike, recoil density, distribution of PKA energy among sub-lattices and number of displaced atoms. The crystallographic features of the zircon structure were found to have profound implications for collision cascades. It was found that energetic PKAs were always deflected into the open channel along the z direction. Their displacements along the longitudinal x direction were never greater than about 4 nm in our simulations. The estimation of the cascade extent assuming homogeneous media thus greatly over-predicts the PKA displacement along the longitudinal direction. The effects of PKA mass on collision cascade were studied by comparing the cascades caused by Zr and U PKAs. The U atoms were simply "super-mass" Zr atoms in this study: U-Zr, U-Si and U-O interactions were the same as Zr-Zr, Zr-Si and Zr-O interactions, respectively. It was found that heavier PKAs produced longer cascades, more structural damage, and higher temperature in thermal spike. U also traveled further along the longitudinal x direction because it was less prone to change of velocity direction. The depleted regions in the core of the cascades surrounded by a densified shell, which were found in simulations by Trachenko et al., were not found in our study. After extensive tests of recently published zircon potentials, it was found that three out of the five tested potentials yielded poor elastic constants and appear to be unfit for serious simulations. Published simulation results using these potentials should accordingly be viewed cautiously.
by Yi Zhang.
Ph.D.
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YeÅ, ilyurt Serhat. "Construction and validation of computer-simulation surrogates for engineering design and optimization". Thesis, Massachusetts Institute of Technology, 1995. http://hdl.handle.net/1721.1/11889.
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Ma, Zhegang. "Development of MURR flux trap model for simulation and prediction of sample loading reactivity worth and isotope production". Diss., Columbia, Mo. : University of Missouri-Columbia, 2007. http://edt.missouri.edu/Winter2007/Dissertation/MaZ-050807-D7038/.
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Thesis (Ph. D.)--University of Missouri-Columbia, 2007.The entire dissertation/thesis text is included in the research.pdf file; the official abstract appears in the short.pdf file (which also appears in the research.pdf); a non-technical general description, or public abstract, appears in the public.pdf file. Title from title screen of research.pdf file (viewed on September 27, 2007) Vita. Includes bibliographical references.
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Mu, Junju. "Computer simulation study of third phase formation in a nuclear extraction process". Thesis, University of Manchester, 2017. https://www.research.manchester.ac.uk/portal/en/theses/computer-simulation-study-of-third-phase-formation-in-a-nuclear-extraction-process(a1ad2143-4fc4-41cf-84c5-e447eeb0b3a3).html.
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Third phase formation is an undesirable phenomenon during the PUREX process, which is a continuous liquid-liquid extraction approach for the reprocessing of uranium and plutonium from spent nuclear fuel. When third phase formation occurs, the organic extraction solution splits into two layers. The light upper layer, which is commonly named the light organic phase, contains a lower concentration of metal ions, tri-n-butyl phosphate (TBP) and nitric acids but is rich in the organic diluent. The heavy lower layer, which is commonly named the third phase, contains high concentrations of metal ions, TBP and nitric acids. As the third phase contains high concentrations of the uranium and plutonium complexes it can thus cause processing and safety concerns. Therefore, a comprehensive understanding of the mechanism of third phase formation is needed so as to improve the PUREX flowsheet. To investigate third phase formation through molecular simulations, one should first obtain reliable molecular models. A refined model for TBP, which uses a new set of partial charges generated from our density functional theory calculations, was proposed in this study. To compare its performance with other available TBP models, molecular dynamics simulations were conducted to calculate the thermodynamic properties, transport properties and the microscopic structures of liquid TBP, TBP/water mixtures and TBP/n-alkane mixtures. To our knowledge, it is only TBP model that has been validated to show a good prediction of the microscopic structure of systems that consist of both hydrophobic and hydrophilic species. This thesis also presents evidence that the light-organic/third phase transition in the TBP/n-dodecane/HNO3/H2O systems, which is relevant to the PUREX process, is an unusual transition between two isotropic, bi-continuous micro-emulsion phases. The light-organic /third phase coexistence was first observed using Gibbs Ensemble Monte Carlo (GEMC) simulations and then validated through Gibbs free energy calculations. Snapshots from the simulations as well as the cluster analysis of the light organic and third phases reveal structures akin to bi-continuous micro-emulsion phases, where the polar species reside within a mesh whose surface consists of amphiphilic TBP molecules. The non-polar n-dodecane molecules are outside this mesh. The large-scale structural differences between the two phases lie solely in the dimensions of the mesh. To our knowledge, the observation of the light-organic/third phase coexistence through simulation approaches and a phase transition of this nature have not previously been reported. Finally, this thesis presents evidence that the microscopic structure of the light organic phase of the Zr(IV)/TBP/n-octane/HNO3/H2O system, which is also related to the PUREX process, is different from that of the common hypothesis, where such system is consisted of large ellipsoidal reverse micelles. Snapshots from simulations, hydrogen bonding analysis and cluster analysis showed that the Zr4+, nitrate, TBP and H2O form extended aggregated networks. Thus, as above, we observe a bi-continuous structure but this time with embedded local clusters centred around the Zr4+ ions. The local clusters were found to consist primarily of Zr(NO3)4·3TBP complexes. This finding provides a new view of the structure of the Zr(IV)/TBP/n-octane/HNO3/H2O system.
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Carlsson, Peter. "Nuclear receptors studied by molecular dynamics computer simulations /". Stockholm, 2004. http://diss.kib.ki.se/2004/91-7349-823-8.
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Anadani, Mohamed. "Decision support systems for nuclear reactor control". Thesis, University of Sheffield, 2000. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.341828.
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Grammatikopoulos, Panagiotis. "Computer simulation of dislocation interaction with radiation-induced obstacles in iron". Thesis, University of Liverpool, 2009. http://livrepository.liverpool.ac.uk/1218/.
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Assessment of candidate materials for fusion power plants provide one of the major structural materials challenges of the next decades. Computer simulation provides a useful alternative to experiments on real-life irradiated materials. Within the framework of a multi-scale modelling approach, atomic scale studies by molecular dynamics (MD) and statics (MS) are of importance, since they enable understanding of atomic interaction mechanisms invisible at coarser scales. Nano-scale defect clusters, such as voids, solute-atom precipitates and dislocation loops can form in metals irradiated by high-energy atomic particles. Since they are obstacles to dislocation glide, they can affect plasticity, substantially changing the yield and flow stresses and ductility. In this study, a model for α-Fe developed by Osetsky and Bacon [26] has been used, that enables dislocation motion under applied shear strain at various temperatures and strain rates. Three main results were obtained. First, the two interatomic potentials used (A97 [79] and A04 [31]) were assessed with respect to reproducing dislocation properties. Both were in good agreement but for one fact: an unexpected and not previously reported displacement of core atoms along the direction of the dislocation line of a 1/2[111](1-10) edge dislocation was observed for the A97 potential. A connection of this phenomenon with differences in Peierls stress values for the two potentials was proposed. Second, the interaction of a 1/2[111](1-10) edge dislocation with a number of different configurations of spherical voids and Cu-precipitates 2 and 4 nm in diameter was investigated. The defects were centred on, above and below the dislocation glide plane. The mechanisms governing the interactions were analysed. For the first time it was observed that by interacting with a void, the dislocation can undergo both positive and negative climb, depending on the void position. A bcc to fcc phase transition was observed for the larger precipitates, in agreement with literature findings. Third, the obstacle strength of 1/2‹111› and ‹100› loops was obtained under various conditions and geometries for both potentials. Interactions are sometimes complex, but could be described in terms of conventional dislocation reactions in which Burgers vector is conserved. The critical resolved shear stress for dislocation breakaway and the fraction of interstitials left behind are wide-ranging. Finally, a mapping of all obstacle strengths was created for the purpose of comparison. ‹100› loops with Burgers vector parallel to the dislocation glide plane and 1/2‹111› loops proved to be strong obstacles. Small size voids are stronger than Cu-precipitates of the same size. The complexity of some reactions and the variety of obstacle strengths poses a challenge for the development of continuum models of dislocation behaviour in irradiated iron.
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Hidalga, García-Bermejo Patricio. "Development and validation of a multi-scale and multi-physics methodology for the safety analysis of fast transients in Light Water Reactors". Doctoral thesis, Universitat Politècnica de València, 2021. http://hdl.handle.net/10251/160135.
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[ES] La tecnología nuclear para el uso civil genera más preocupación por la seguridad que muchas otras tecnologías que se usan a diario. La Autoridad Nuclear define las bases de cómo debe realizarse la operación segura de una Central Nuclear. De acuerdo a las directrices establecidas por la Autoridad Nuclear, una Central Nuclear debe analizar una envolvente de escenarios hipotéticos y comprobar de manera determinista que los criterios de aceptación para dicho evento se cumplen. El Análisis Determinista de Seguridad utiliza herramientas de simulación que aplican la física conocida sobre el comportamiento de la Central Nuclear para evaluar la evolución de una variable de seguridad y asegurar que los límites no se sobrepasan.
El desarrollo de la tecnología informática, de los métodos matemáticos y de la física que envuelve el comportamiento de una Central Nuclear han proporcionado herra-mientas de simulación potentes que son capaces de predecir el comportamiento de las variables de seguridad con una importante precisión. Esto permite analizar escenarios de manera más realista evitando asumir condiciones conservadoras que hasta la fecha compensaban la falta de conocimiento modelado en las herramientas de simulación. Las herramientas conocidas como De Mejor Estimación son capaces de analizar even-tos transitorios en diferentes escalas. Además, emplean modelos analíticos de las dife-rentes físicas más detallados, así como correlaciones experimentales más realistas y actuales. Un paso adelante en el Análisis Determinista de Seguridad pretende combinar las diferentes herramientas de Mejor Estimación que se emplean para analizar las dis-tintas físicas de una Central Nuclear, considerando incluso la interacción entre ellas y el análisis progresivo a diferentes escalas, llegando a analizar fenómenos más locales si es necesario.
Para este fin, esta tesis presenta una metodología de análisis multi-físico y multi-escala que emplea diferentes códigos de simulación analizando el escenario propuesto a dife-rentes escalas, es decir, desde un nivel de planta que incluye los distintos componentes, hasta el volumen de control que supone el refrigerante pasando entre las varillas de combustible. Esta metodología permite un flujo de información que va desde el análi-sis a mayor escala hasta el de menor escala. El desarrollo de esta metodología ha sido validado con datos de planta para poder evaluar el alcance de esta metodología y pro-porcionar nuevas líneas de trabajo futuro. Además, se han añadido los resultados de los distintos procesos de validación y verificación que han surgido a lo largo de este trabajo.[CA] La tecnologia nuclear per a l'ús civil genera més preocupació per la seguretat que moltes altres tecnologies d'ús quotidià. L'Autoritat Nuclear defineix les bases de com ha de realitzar-se l'operació segura d'una Central Nuclear. D'acord amb les directrius establertes per l'Autoritat Nuclear, una Central Nuclear ha d'analitzar una envoltant d'escenaris hipotètics I comprovar de manera determinista que els criteris d'acceptació per a l'esdeveniment seleccionat es compleixen. L'Anàlisi Determinista de Seguretat utilitza eines de simulació que apliquen la física coneguda sobre el comportament de la Central Nuclear per avaluar l'evolució d'una variable de seguretat i assegurar que els límits no es traspassen. El desenvolupament de la tecnologia informàtica, els mètodes matemàtics i de la física que envolta el comportament d'una Central Nuclear han proporcionat eines de simulació potents amb capacitat de predir el comportament de les variables de seguretat amb una precisió significativa. Això permet analitzar escenaris de manera realista evitant assumir condicions conservadores que fins al moment compensaven la mancança de coneixement. Les eines de simulació conegudes com De Millor Estimació son capaces d'analitzar esdeveniment transitoris a diferent escales. A més, utilitzen models analítics per a les diferents físiques amb més detall així com correlacions experimentals més actualitzades i realistes. Un pas més endavant en l'Anàlisi Determinista de Seguretat pretén combinar les diferents eines de Millor Estimació que se utilitzen per analitzar les distintes físiques d'una Central Nuclear, considerant inclús la interacció entre ells i l'anàlisi progressiu a diferents escales, amb la finalitat de poder analitzar fenòmens locals. Per a aquest fi, esta tesi presenta una metodologia d'anàlisi multi-física i multi-escala que utilitza diferents codis de simulació analitzant l'escenari proposat a diferents escales, és a dir, des d'un nivell de planta que inclou els distints components, fins al volum de control que suposa el refrigerant passant entre les varetes de combustible. Esta metodologia permet un flux de informació que va des de l'anàlisi d'una escala major a una menor. El desenvolupament d'aquesta metodologia ha sigut validada i verificada amb dades de planta i els resultats han sigut analitzats a fi d'avaluar la capacitat de la metodologia i les possibles línies de treball futur. A més s'han afegit els principals resultats de verificació i validació que han sorgit en les distintes etapes d'aquest treball.
[EN] The nuclear technology for civil use has generated more concerns for the safety than several other technologies applied to the daily life. The Nuclear Regulators define the basis of how the Safety Operation of Nuclear Power Plants is to be done. According to these guidelines, a Nuclear Power Plant must analyze an envelope of hypothetical events and deterministically define if the acceptance criteria for these events is met. The Deterministic Safety Analysis uses simulation tools that apply the physics known in the behavior of the Nuclear Power Plant to evaluate the evolution of a safety varia-ble and assure that the safety limits will not be exceeded. The development of the computer science, the numerical methods and the physics involved in the behavior of a Nuclear Power Plant have yield powerful simulation tools that are capable to predict the evolution of safety variables which significant accuracy. This allows to consider more realistic simulation scenarios instead of con-servative approaches in order to compensate the lack of knowledge in the applied prediction methods. The so called Best Estimate simulation tools are capable to analyze the transient events in different scales. Furthermore, they account more detailed analytical models and experimental correlations. A step forward in the Deterministic Safety Analysis intends to combine the Best Estimate simulation tools of the different physics considering the interaction among them and analyzing the different scales, considering more local approaches if necessary. For this purpose, this thesis work presents a multi-scale and multi-physics methodology that uses different physics codes and has the aim of modeling postulated scenarios in different scales, i.e. from system models representing the components of the plants to the subchannel models that analyze the behavior of the coolant between the fuel rods. This methodology allows a flow of information where the output of one scale is used as input in a more detailed scale to predict a more local analysis of parameters, such as the Critical Power Ratio, which are of great importance for the estimation of safety margins. The development of this methodology has been validated against plant data with the aim of evaluating the scope of this methodology and in order to provide future lines of development. In addition, different results of the validation and verifi-cation yielded in the development of the parts of this methodology are presented.
Hidalga García-Bermejo, P. (2020). Development and validation of a multi-scale and multi-physics methodology for the safety analysis of fast transients in Light Water Reactors [Tesis doctoral]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/160135
TESIS
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Zhan, Yiyi. "PC-based visual simulation of high pressure arc plasma". Thesis, University of Liverpool, 2011. http://livrepository.liverpool.ac.uk/3433/.
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Chu, Kwun Pok. "Computational studies of nuclear receptors : estrogen receptors, glucocorticoid receptors, and farnesoid X receptor". HKBU Institutional Repository, 2009. http://repository.hkbu.edu.hk/etd_ra/1058.
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Tran, Trung Nam. "Surface discharge dynamics : theory, experiment and simulation". Thesis, University of Southampton, 2010. https://eprints.soton.ac.uk/165509/.
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The use of solid insulators in electrical generation, transmission and distribution is widespread. However, the accumulation of charge on the insulator surface has proved to be one of the major factors contributing to system failures. This research work is aimed at studying the dynamics of surface discharge in theory, by simulation and experiment. Different surface charging theories have been reviewed and classiffied according to electric field uniformity. The focus is on basic processes involved in the formation of positive and negative surface discharges. The experimental work utilises the non-destructive quantitative Pockels technique to measure surface charge density distribution. Practical considerations of the Pockels experiment together with image processing techniques are discussed in detail. Using this technique, various factors which influence the surface discharge dynamics have been studied including the effects of the applied voltage waveform, electrode shape and local gaseous environment. Results obtained using positive/negative square wave, ramp and sinusoidal voltages are reported. The impact of using a mushroom electrode instead of a needle electrode is also analysed. In addition, various insulation gases have been experimented namely dry air, N2, CO2 and their mixtures with SF6. Surface discharge measurements have been performed in these gases at various levels of pressure. Surface discharge modelling and simulation studies have also been undertaken. The simulation principles are based on a system of coupled hydrodynamic equations consisting of continuity and Poisson's equations. By solving these equations, the movement and interaction of charged particles and transient electric eld can be simulated and used to verify the discharge theories and experimental results. Due to the asymmetric lamentary nature of positive surface streamers, the development of a positive surface discharge is separated into two phases. The rst phase involves the axial streamer development in the gas gap between the needle electrode and the dielectric surface. This phase is simulated in 2D axial symmetry space dimension by the nite element package COM-SOL. The second phase simulates the streamer propagation in 1D along the dielectric surface by using the eld results from the rst phase. This part of the model is solved by the accurate ux-corrected transport algorithm. The effects of model parameters on the simulation results are discussed and a comparison with experimental data made. Prior to the simulation of a negative surface discharge, a negative corona discharge model in 2D axial symmetry has been analysed (Trichel pulses). The model behaviour is studied with reference to experimental data as model parameters are varied. When the insulators are introduced, the accumulation of surface charge distorts the electric eld leading to the formation of only one discharge current pulse. The simulation charge density distribution is in good agreement with results obtained from the Pockels experiment.
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Cheng, Kit-yan Ruby y 鄭潔茵. "Nuclear emergency preparedness model based on Daya Bay Nuclear Power stations for educational purposes". Thesis, The University of Hong Kong (Pokfulam, Hong Kong), 2005. http://hub.hku.hk/bib/B36168464.
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PALADINO, PATRICIA A. "Pré-processador matemático para o código RELAP5 utilizando o Microsoft Excel". reponame:Repositório Institucional do IPEN, 2006. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11403.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Valiullin, Rustem, Jörg Kärger y Peter Monson. "Adsorption hysteresis in nanopores: options to a comparative study using nuclear magnetic resonance and computer simulation methods". Diffusion fundamentals 2 (2005) 107, S. 1-2, 2005. https://ul.qucosa.de/id/qucosa%3A14445.
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Mattos, Carlos Eduardo. "Estudo de modelos para o comportamento a altas queimas de varetas combustível urânio - 7% gadolínio para reatores a água leve pressurizada: avaliação dos parâmetros para prolongamento do tempo de queima do núcleo". Universidade de São Paulo, 2018. http://www.teses.usp.br/teses/disponiveis/85/85134/tde-17052018-160542/.
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O objetivo deste trabalho é verificar os resultados fornecidos pelo programa computacional FRAPCON-3, hoje na versão 5, utilizado no processo de simulação do comportamento de varetas combustíveis de reatores refrigerados a água pressurizada (Pressurized Water Reactor PWR), sob situações operacionais de regime permanente, em condições de alta queima. Para realizar a verificação, foi utilizada a base de dados FUMEX-III, que fornece dados relativos a experimentos realizados com diversos tipos de combustíveis nucleares, submetidos a diversas condições operacionais. Através dos resultados obtidos na simulação do programa FRAPCON-3.5 e da sua comparação com os dados experimentais da base FUMEX-III, foi possível constar que o programa possui boa capacidade de predizer o comportamento operacional da vareta combustível em regime permanente a altas queimas. O trabalho consiste também em verificar a correlação entre UO2 e UO2-7%Gd2O3 na análise dos modelos que simulam o comportamento das pastilhas combustível. A adição do óxido de gadolínio ou gadolínia (Gd2O3), constitui-se na opção tecnológica mais solidamente consagrada e hoje comum em várias centrais nucleares. Por meio dos resultados obtidos nas simulações computacionais foram apresentadas e discutidas a influência das propriedades do UO2 e UO2-7%Gd2O3, quanto à temperatura no centro do combustível, liberação de gás de fissão na vareta, temperatura média do revestimento, volume interno e pressão interna da vareta combustível.The objective of this work is to verify the results provided by the computer program FRAPCON-3, now in version 5, used in the simulation process of the behavior of fuel rods of pressurized water reactors - PWR permanent, in conditions of high burn. In order to carry out the verification, the FUMEX-III database was used, which provides data on experiments performed with different types of nuclear fuel, under different operating conditions. The results obtained in the simulation of the FRAPCON-3.5 program and its comparison with the experimental data of the FUMEX-III base showed that the program has a good ability to predict the operational behavior of the fuel rod in a steady state at high burn. The work also consists in verifying the correlation between UO2 and UO2-7%Gd2O3 in the analysis of models that simulate the behavior of fuel pellets. The addition of gadolinium oxide (Gd2O3) constitutes the most solidly established and now common technological option in several nuclear power plants. The influence of the properties of UO2 and UO2-7%Gd2O3 on the temperature at the center of the fuel, fission gas release on the rod, average coating temperature, internal volume and pressure were presented and discussed. of the fuel rod.
35
REIS, REGIS. "Análise do comportamento sob irradiação do combustível nuclear a altas queimas com os programas computacionais FRAPCON e FRAPTRAN". reponame:Repositório Institucional do IPEN, 2014. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11797.
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Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
36
Schiepp, Thomas. "A simulation method for design and development of magnetic shape memory actuators". Thesis, University of Gloucestershire, 2015. http://eprints.glos.ac.uk/2974/.
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The systems/products and their design processes have become more and more complicated due to the fact that their requirements in terms of function, durability, reliability and energy efficiency have been increased significantly and that their leading time has to be short and their materials cost has to be low. To meet these requirements, individual parts and subsystems have to offer increased functionality and efficiency themselves. It has been found that smart materials, such as piezo ceramics or various shape memory alloys as well as less known dielectric elastomers or magnetic shape memory alloys, offer ideal preconditions to fulfil such requirements. Among the various shape memory alloys, the Magnetic Shape Memory (MSM) alloy is a kind of smart material that can elongate and contract in a magnetic field. Based on the MSM alloy a new type of smart electromagnetic actuators have been designed and developed. This kind of actuator exhibits the features above. Typically, the MSM material is a monocrystalline Ni-Mn-Ga alloy, which has the ability to change its size or shape very fast and many million times repeatedly. State-of-the-art alloys are able to achieve a magnetic field induced strain of up to 12%. The magneto-mechanical characteristic of MSM alloys is being constantly improved. However, as far as the author is aware, there are no efficient and commercially available tools for engineers to design MSM-based actuators. To achieve this, simulation tools for design are indispensable. This thesis is dedicated to this task. In this PhD thesis, new design and simulation techniques for MSM-based actuators have been studied. In particular, three simulation methods have been proposed. These three methods extend standard magneto-static FEM simulation techniques by taking into account the magneto-mechanical coupling and the magnetic anisotropy of the MSM materials. They differ in terms of the necessary a priori alloy characterisation (i.e., measurement effort), computational complexity and consequent computing time. The magneto-mechanical characteristics of the MSM material are a necessary and fundamental ingredient for this type of simulation. However, the characterisation of the MSM materials is a very challenging task and requires specific modifications to standard measurement approaches. So, in this thesis, some specific measurement methods of the magneto-mechanical characteristics of the MSM materials have been proposed, designed and developed. It is described how existing measurement instruments can be modified to measure the unique magneto-mechanical characteristics of MSM, so they are applicable and with practical values. Various tests have been carried out to validate the new methods and the necessary characterisations of the properties of MSM materials have been performed, such as the measurement of the permeability of MSM under a defined stress during elongation. The new measurement results have been analysed and the findings have been used to design and develop the simulation methods. The three simulation methods can be used to predict and optimise the current-elongation behaviour of an MSM element under the load of a mechanical stress while excited by a magnetic field. Extensive experiments have been carried out to validate these three simulation methods. The results show that the three methods are relatively simple but, at the same time, very effective means to model, predict and optimise the properties of an MSM actuator using finite element tools. In addition, the experiment results have also shown that the simulation methods can be used to gain some deep insights into the magneto-mechanical interaction between the MSM element and the electromagnetic actuator. In this thesis an evolutionary algorithm which works together with the simulation methods has been developed to achieve individual optimised solutions in very short times. In summary, from the experiment results, it has been found that the measurements and simulation methods proposed and developed in this thesis; enable designers to perform simulations for a high-quality actuator design based on the magneto-mechanical properties of MSM alloys. This is the first time that a MSM can be characterised for simulation purposes in a fast and precise way to predict MSM and electromagnetic actuator interactions and identify and optimise the design parameters of such actuators. However, these simulation methods are strongly dependent on the measurement of the magneto-mechanical characteristics of magnetic shape memory alloys, whose precision can be further improved. To reach commercial success as well higher precision in the simulation prediction, further achievements in the field of material science (e.g. smoothness of mechanical curves) are also necessary.
37
Havet, Maxime. "Solution of algebraic problems arising in nuclear reactor core simulations using Jacobi-Davidson and multigrid methods". Doctoral thesis, Universite Libre de Bruxelles, 2008. http://hdl.handle.net/2013/ULB-DIPOT:oai:dipot.ulb.ac.be:2013/210467.
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The solution of large and sparse eigenvalue problems arising from the discretization of the diffusion equation is considered. The multigroupdiffusion equation is discretized by means of the Nodal expansion Method (NEM) [9, 10]. A new formulation of the higher order NEM variants revealing the true nature of the problem, that is, a generalized eigenvalue problem, is proposed. These generalized eigenvalue problems are solved using the Jacobi-Davidson (JD) method
[26]. The most expensive part of the method consists of solving a linear system referred to as correction equation. It is solved using Krylov subspace methods in combination with aggregation-based Algebraic Multigrid (AMG) techniques. In that context, a particular
aggregation technique used in combination with classical smoothers, referred to as oblique geometric coarsening, has been derived. Its particularity is that it aggregates unknowns that
are not coupled, which has never been done to our
knowledge. A modular code, combining JD with an AMG preconditioner, has been developed. The code comes with many options, that have been tested. In particular, the instability of the Rayleigh-Ritz [33] acceleration procedure in the non-symmetric case has been underlined. Our code has also been compared to an industrial code extracted from ARTEMIS.
Doctorat en Sciences de l'ingénieur
info:eu-repo/semantics/nonPublished
38
Edwards, Luke J. "Highly efficient quantum spin dynamics simulation algorithms". Thesis, University of Oxford, 2014. http://ora.ox.ac.uk/objects/uuid:3eec480e-5a3a-4197-a786-e6d42988d4a5.
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Spin dynamics simulations are used to gain insight into important magnetic resonance experiments in the fields of chemistry, biochemistry, and physics. Presented in this thesis are investigations into how to accelerate these simulations by making them more efficient. Chapter 1 gives a brief introduction to the methods of spin dynamics simulation used in the rest of the thesis. The `exponential scaling problem' that formally limits the size of spin system that can be simulated is described. Chapter 2 provides a summary of methods that have been developed to overcome the exponential scaling problem in liquid state magnetic resonance. The possibility of utilizing the multiple processors prevalent in modern computers to accelerate spin dynamics simulations provides the impetus for the investigation found in Chapter 3. A number of different methods of parallelization leading to acceleration of spin dynamics simulations are derived and discussed. It is often the case that the parameters defining a spin system are time-dependent. This complicates the simulation of the spin dynamics of the system. Chapter 4 presents a method of simplifying such simulations by mapping the spin dynamics into a larger state space. This method is applied to simulations incorporating mechanical spinning of the sample with powder averaging. In Chapter 5, implementations of several magnetic resonance experiments are detailed. In so doing, use of techniques developed in Chapters 2 and 3 are exemplified. Further, specific details of these experiments are utilized to increase the efficiency of their simulation.
39
Holt, Lars [Verfasser], Rafael [Akademischer Betreuer] [Gutachter] Macián-Juan y HANS-JOSEF [Gutachter] ALLELEIN. "Improvement and validation of a computer model for the thermo-mechanical fuel rod behavior during reactivity transients in nuclear reactors / Lars Holt ; Gutachter: Hans-Josef Allelein, Rafael Macián-Juan ; Betreuer: Rafael Macián-Juan". München : Universitätsbibliothek der TU München, 2017. http://d-nb.info/1147565600/34.
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Rader, Jordan D. "Loss of normal feedwater ATWS for Vogtle Electric Generating Plant using RETRAN-02". Thesis, Atlanta, Ga. : Georgia Institute of Technology, 2009. http://hdl.handle.net/1853/31741.
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Thesis (M. S.)--Nuclear Engineering, Georgia Institute of Technology, 2010.Committee Chair: Abdel-Khalik, Said I.; Committee Member: Ghiaasiaan, S. Mostafa; Committee Member: Hertel, Nolan E. Part of the SMARTech Electronic Thesis and Dissertation Collection.
41
Dahlfors, Marcus. "Studies of Accelerator-Driven Systems for Transmutation of Nuclear Waste". Doctoral thesis, Uppsala : Acta Universitatis Upsaliensis : Universitetsbiblioteket [distributör], 2006. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-6341.
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SANTOS, DIOGO F. dos. "Caracterização dos campos neutrônicos obtidos por meio de armadilhas de nêutrons a partir da utilização de água pesada (D2O) no interior do núcleo do reator nuclear IPEN/MB-01". reponame:Repositório Institucional do IPEN, 2015. http://repositorio.ipen.br:8080/xmlui/handle/123456789/23825.
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Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
43
MASSICANO, FELIPE. "Modelagem de um sistema de planejamento em radioterapia e medicina nuclear com o uso do código MCNP6". reponame:Repositório Institucional do IPEN, 2015. http://repositorio.ipen.br:8080/xmlui/handle/123456789/26371.
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Tese (Doutorado em Tecnologia Nuclear)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
44
Galeazzo, Flavio Cesar Cunha. "Modelagem de um reator com serpentinas axiais utilizando a fluido dinamica computacional - CFD". [s.n.], 2005. http://repositorio.unicamp.br/jspui/handle/REPOSIP/267348.
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Orientador: Jose Roberto NunhezDissertação (mestrado) - Universidade Estadual de Campinas, Faculdade de Engenharia Quimica
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Resumo: O trabalho analisou o impacto que alterações nas condições operacionais e na geometria causam no desempenho térmico de um reator de esterificação de porte industrial da empresa M&G. As análises utilizaram um modelo numérico criado com o software FLUENT. A hipótese simplificadora de se considerar somente uma fase líquida, apesar do sistema reacional real ser trifásico, foi competente em mostrar as principais características do escoamento e da transferência de calor no interior do reator, como zonas de estagnação e área de atuação dos impelidores, além da troca de calor a partir dos tubos da serpentina axial. Os resultados levam à conclusão de que o sistema de agitação possui um papel apenas secundário no desempenho térmico do reator, frente o papel preponderante dos discos separadores de fluxo
Abstract: The work analysed the impact of operating conditions and geometric changes in the thermal performance of a esterification industrial reactor of the company M&G. The analyses used a numerical model created with the aid of the software FLUENT. The simplifying hypothesis of considering only one liquid phase instead of the real system three phases was capable of showing the main characteristics of the flow and the heat transfer inside the reactor, like stagnation zones and impellers influence zones, and the heat transfer from the axial coil. The results lead to the conclusion that the agitation system have only a secundary role in the reactor thermal performance, while the role of the flow separation discs is dominant
Mestrado
Desenvolvimento de Processos Químicos
Mestre em Engenharia Química
45
Manalo, Kevin. "Detailed analysis of phase space effects in fuel burnup/depletion for PWR assembly & full core models using large-scale parallel computation". Diss., Georgia Institute of Technology, 2013. http://hdl.handle.net/1853/50351.
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Nuclear nonproliferation research and forensics have a need for improved software solutions, particularly in the estimates of the transmutation of nuclear fuel during burnup and depletion. At the same time, parallel computers have become effectively sized to enable full core simulations using highly-detailed 3d mesh models. In this work, the capability for modeling 3d reactor models is researched with PENBURN, a burnup/depletion code that couples to the PENTRAN Parallel Sn Transport Solver and also to the Monte Carlo solver MCNP5 using the multigroup option. This research is computationally focused, but will also compare a subset of results of experimental Pressurized Water Reactor (PWR) burnup spectroscopy data available with a designated BR3 PWR burnup benchmark. Also, this research will analyze large-scale Cartesian mesh models that can be feasibly modeled for 3d burnup, as well as investigate the improvement of finite differencing schemes used in parallel discrete ordinates transport with PENTRAN, in order to optimize runtimes for full core transport simulation, and provide comparative results with Monte Carlo simulations. Also, the research will consider improvements to software that will be parallelized, further improving large model simulation using hybrid OpenMP-MPI. The core simulations that form the basis of this research, utilizing discrete ordinates methods and Monte Carlo methods to drive time and space dependent isotopic reactor production using the PENBURN code, will provide more accurate detail of fuel compositions that can benefit nuclear safety, fuel management, non-proliferation, and safeguards applications.
46
Lucca, Eneida Aparecida de. "Modelagem e simulação de reatores industriais em fase liquida do tipo Loop para polimerização de propileno". [s.n.], 2007. http://repositorio.unicamp.br/jspui/handle/REPOSIP/266238.
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Orientadores: Rubens Maciel Filho, Jose Carlos Costa da Silva Pinto, Priamo Albuquerque Melo JuniorDissertação (mestrado) - Universidade Estadual de Campinas, Faculdade de Engenharia Química
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Resumo: Reatores tubulares do tipo loop são amplamente empregados nas indústrias de poliolefinas. No caso da produção de polipropileno, compõem a tecnologia Spheripol. São constituídos de duas seções tubulares interconectadas por um ponto de alimentação e por uma bomba, que promove a recirculação da massa reacional. O simulador dinâmico desenvolvido nesse trabalho é capaz de estimar valores de diversas variáveis chave no monitoramento do processo; dentre elas, o XS e o MFI. As validações feitas mostraram que o simulador é capaz de representar de forma acurada os dados experimentais disponíveis em uma planta real de polimerização, inclusive para ¿N¿ reatores em série
Abstract: Tubular loop reactors are widely used in the polyolefins industries. In the particular case of polypropylene production, loop reactors are part of the Spheripol technology. Loop reactors are composed of two tubular reactors that are connected by a feed point and a pump that is responsible for promoting recirculation of the reaction mass. The dynamic simulator developed here is able to estimate values of several important variables used to monitor the industrial process, like the XS and the MFI. The model was validated with actual industrial data obtained for different reactor configurations, including ¿N¿ reactors in series
Mestrado
Engenharia de Processos
Mestre em Engenharia Química
47
Spogis, Nicolas. "Desenvolvimento de um impelidor de alta eficiência através da dinâmica dos fluídos computacional e otimização multi-objetivo". [s.n.], 2007. http://repositorio.unicamp.br/jspui/handle/REPOSIP/266878.
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Orientador: José Roberto NunhezTese (doutorado) - Universidade Estadual de Campinas, Faculdade de Engenharia Química
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Resumo: Atualmente o ciclo de desenvolvimento de produtos requer um baixo tempo de projeto e mínimos custos. Ao mesmo tempo, a qualidade final do produto não deve ser afetada; pelo contrário, as companhias precisam melhorar seus produtos para se manterem no mercado competitivo. Em muitas indústrias, o uso de software de otimização de projetos está se tornando a principal ferramenta para alcançar rapidamente estas metas aparentemente contraditórias. O projeto de processos químicos e equipamentos é uma tarefa que exige um apoio experimental significante e um grande número de protótipos e testes. Visando reduzir o tempo de desenvolvimento, os softwares ANSYS CFX e modeFRONTIER foram acoplados a fim de obter um projeto de um impelidor de alta eficiência para aplicações "flow-controlled" e para mistura de produtos de baixa viscosidade. A análise de desempenho do impelidor foi realizada através do modelo de turbulência SST (Shear- Stress Transport) acoplado com um modelo de correção para curvatura das linhas de correntes. O modelo SST combina as vantagens dos modelos k- ? e k-?, garantindo uma excelente relação entre a tensão turbulenta e energia cinética turbulenta, além de fornecer uma predição precisa e robusta de descolamentos/separações da camada limite. Os modelos "Multiple Frames of Reference" e "Frozen Rotor Frame Change" foram usados para investigar a interação entre as partes móveis (impelidor) e as partes estáticas (parede do vaso e chicanas) no tanque de mistura. Um algoritmo estocástico robusto (MOGA II) foi utilizado como método de otimização. A otimização multi-objetivo possui sete variáveis de entrada, duas restrições não lineares, e duas funçõesobjetivo. Através deste estudo foi possível obter um aumento simultâneo da capacidade de bombeamento do impelidor e da homogeneidade de mistura
Abstract: Nowadays product development cycle requires shorter turnaround times and lowers costs. At the same time, quality should not suffer; on the contrary, companies need to improve their products in order to remain competitive. In many industries, the use of design optimization software is fast becoming the major way to achieve these apparently conflicting goals. The project of chemical processes and equipments is a task that demands a significant experimental support and a great number of prototypes and tests. Aiming at reducing the development time, ANSYS CFX tools have been successfully coupled to modeFRONTIER so as to lead to an optimal design of a high efficiency impeller for flow-controlled, low viscosity applications. The analysis of impeller shape performance was carried out with the SST (Shear-Stress Transport) model coupled with the streamline curvature turbulence model. The SST model combines the advantages from the k-? and k-? models, ensuring proper relation between turbulent stress and turbulent kinetic energy and allowing accurate and robust prediction of the impeller blade flow separation. The Multiple Frames of Reference and the Frozen Rotor Frame Change model were used in order to investigate the rotor/stator interaction inside the mixing vessel. A robust stochastic algorithm was used for the automatic multi-objective constrained shape design process. The multi-objective function has seven design variables, two nonlinear constraints, and two objective functions. Simultaneous increase of the impeller pumping capacity and mixing vessel homogeneity were achieved using this method
Doutorado
Desenvolvimento de Processos Químicos
Doutor em Engenharia Química
48
Govers, Kevin. "Atomic scale simulations of noble gases behaviour in uranium dioxide". Doctoral thesis, Universite Libre de Bruxelles, 2008. http://hdl.handle.net/2013/ULB-DIPOT:oai:dipot.ulb.ac.be:2013/210509.
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Nuclear fuel performance is highly affected by the behaviour of fission gases, particularlyat elevated burnups, where large amounts of gas are produced and can
potentially be released. The importance of fission gas release was the motivation
for large efforts, both experimentally and theoretically, in order to increase our
understanding of the different steps of the process, and to continuously improve
our models.
Extensions to higher burnups, together with the growing interest in novel types
of fuels such as inert matrix fuels envisaged for the transmutation of minor actinides,
make that one is still looking for a permanently better modelling, based
on a physical understanding and description of all stages of the release mechanism.
Computer simulations are nowadays envisaged in order to provide a better
description and understanding of atomic-scale processes such as diffusion, but even
in order to gain insight on specific processes that are inaccessible by experimental
means, such as the fuel behaviour during thermal spikes.
In the present work simulation techniques based on empirical potentials have
been used, focusing in a first stage on pure uranium dioxide. The behaviour of
point defects was at the core of this part, but also the estimation of elastic and
melting properties.
Then, in a second stage, the study has been extended to the behaviour of helium
and xenon. For helium, the diffusion in different domains of stoichiometry
was considered. The simulations enabled to determine the diffusion coefficient and
the migration mechanism, using both molecular dynamics and static calculation
techniques. Xenon behaviour has been investigated with the additional intention
to model the behaviour of small intragranular bubbles, particularly their interaction
with thermal spikes accompanying the recoil of fission fragments. For that
purpose, a simplified description of these events has been proposed, which opens
perspectives for further work.
/
Les performances du combustible nucléaire sont fortement affectées par le comportement
des gaz de fission, et ce particulièrement lorsqu’un taux d’épuisement
élevé est atteint, puisque d’importantes quantités de gaz sont alors produites
et peuvent potentiellement être relâchées. Les enjeux, entre autre économiques,
liés au relâchement de gaz de fission ont donné lieu à d’importants efforts, tant
sur le plan expérimental que théorique, afin d’accroître notre compréhension des
différentes étapes du processus, et d’améliorer sans cesse les mod`eles. Les extensions
à des taux d’épuisements encore plus élevés ainsi que l’intérêt croissant pour
de nouveaux types de combustible tels que les matrices inertes, envisages en vue
de la transmutation des actinides mineures, font qu’à l’heure actuelle, le besoin
permanent d’une meilleure modélisation, basée sur une compréhension et une description
physique des différentes étapes du processus de relâchement de gaz de
fission, est toujours de mise.
Les simulations par ordinateur ont ainsi été considérée comme un nouvel angle
de recherche sur les processus élémentaires se produisant à l’échelle atomique, à la
fois afin d’obtenir une meilleure compréhension de processus tels que la diffusion
atomique ;mais aussi afin d’avoir accès à certains processus qui ne sont pas observables
par des voies expérimentales, tels que la le comportement du combustible
lors de pointes thermiques.
Dans ce travail, deux techniques, basées sur l’utilisation de potentiels interatomiques
empiriques, ont permis d’étudier le dioxyde d’uranium, dans un premier
temps en l’absence d’impuretés. Cette partie était principalement centrée sur le
comportement des défauts ponctuels, mais a aussi concerné différentes propriétés
élastiques, ainsi que le processus de fusion du composé.
Ensuite l’étude a été étendue aux comportements de l’hélium de du xénon. Pour
ce qui a trait à l’hélium, la diffusion dans différents domaines de stoechiométrie
a été considérée. Les simulations ont permis de déterminer le coefficient de diffusion
ainsi que le mécanisme de migration lui-même. Quant au xénon, outre les
propriétés de diffusion, l’intention fut de se diriger vers la modélisation des petites
bulles intragranulaires, et plus précisément vers leur interaction avec les pointes
thermiques, créées lors du recul des fragments de fission. Une description simplifiée de ce processus a été proposée, qui offre de nouvelles perspectives dans ce
domaine.
Doctorat en Sciences de l'ingénieur
info:eu-repo/semantics/nonPublished
49
Mendoza, Marin Florentino Lazaro. "Modelagem, simulação e analise de desempenho de reatores tubulares de polimerização com deflectores angulares internos". [s.n.], 2004. http://repositorio.unicamp.br/jspui/handle/REPOSIP/267665.
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Orientadores: Rubens Maciel Filho, Liliane Maria Ferrareso LonaTese (doutorado) - Universidade Estadual de Campinas, Faculdade de Engenharia Quimica
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Resumo: O modelo determinístico e processo homopolimerização na emulsão do estireno são aplicados em reator tubular contínuo sem e com deflectores angulares internos sob condição isotérmica e não isotérmica. Os resultados de modelagem e simulação foram realizados a estado estacionário, modelo unidimensional, coordenada cilíndrica, fluxo pistão laminar completamente desenvolvido, modelo Smith-Ewart para estimar a conversão do monômero, cinética química de Arrhenius corno modelo de velocidade finita laminar para computar a geração química. O objetivo é modelar, simular e analisar o comportamento do reator de homopolimerização na emulsão do estireno com deflectores angulares inclinados internos, e comparar com reator tubular. Os métodos experimental e matemático-dedutivo foram aplicados para obter resultados, por meio de programação computacional, usando Dinâmica de Fluido Computacional através do método de volumes finitos. As seguintes variáveis como temperatura de reação constante e variável, reator tubular sem e com deflectores, temperatura de alimentação, diâmetro de reator, processo adiabático e exotérmico, calor de reação constante e velocidade axial completamente desenvolvida foram investigados. Os efeitos de conversão de monômero, área transversal interna, temperatura axial, concentração do polímero, radicais e iniciador, outros corno densidade de polímero e monômero, perda de carga e queda de pressão foram determinados e simulados. Os produtos foram caracterizados com Número de Partículas (nucleação homogênea e heterogênea), distribuição de peso molecular, tamanho de partículas de polímero e distribuição de viscosidade. Estes resultados foram validados com resultados da literatura sob condição igualou aproximada. Os resultados sob condições não isotérmicas foram melhores que os resultados isotérmicos em termos de caracterização do polímero. Isso mostra que o desenho alternativo proposto (com deflectores) permite obter o polímero com propriedades melhores em termos de número de partículas, distribuição de peso molecular, distribuição do tamanho de partículas e viscosidade
Abstract: Deterministic model and emulsion homopolymerization process of styrene are applied in continuous tubular reactor without and with internal angular baffles under isothermic and no isothermic conditions. The modeling and simulation results were approximate to steady state, one-dimensional model, cylindrical coordinate, fully developed laminar plug flow, Smith-Ewart model to estimate the monomer conversion, Arrhenius chemical kinetics as laminar finite-rate model to compute chemical source. The objective is to model, simulate and to analyze the emulsion homopolymerization reactor performance of styrene with internal-inc1ined angular baffles, and to compare with continuous tubular reactor. The experimental and mathematical-deductive methods were applied to obtain results, by means of computational programming, using Computational Fluid Dynamics (program code), finite volume method. The following variables such as constant and variable reaction temperature, tubular reactor without and with baffles, feed temperature, reactor diameter, adiabatic and exothermic process, constant reaction heat and fully developed axial velocity were investigated. The monomer conversion, internal transversal are a, axial temperature, concentration of polymer, radicals and initiator, others as density of polymer and monomer, head loss and pressure drop effects were determined and simulated. The products were characterized by partic1es number (homogeneous and heterogeneous nuc1eation), molecular weight distribution, polymer partic1es size and polymer viscosity distribution. These results were validated with literature results under same or approximate condition. The results under no isothermic conditions were better than isothermic results in terms of polymer characterization. It is shown that the proposed alternative design (with baffles) allow to obtain the polymer with better properties in terms of number of partic1es, molecular weight distribution, particle size distribution and viscosity
Doutorado
Desenvolvimento de Processos Químicos
Mestre em Engenharia Química
50
Bastos, Jaci Carlo Schramm Camara. "Simulação do escoamento gas-solido em um duto cilindrico vertical em leito fluidizado rapido aplicando a tecnica CFD". [s.n.], 2005. http://repositorio.unicamp.br/jspui/handle/REPOSIP/266327.
Texto completoResumen
Orientador: Milton MoriDissertação (mestrado) - Universidade Estadual de Campinas, Faculdade de Engenharia Quimica
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Resumo: A presente pesquisa apresenta a modelagem matemática e a simulação de uma operação de fluidização rápida em um longo duto cilíndrico vertical, operação importante em vários processos industriais, sendo que sua principal aplicação está vinculada ao craqueamento catalítico do petróleo para a conversão em gasolina. Tem como objetivos a obtenção de uma compreensão contínua e o conhecimento do desenvolvimento do escoamento gás-sólido, bem como a soma de experiências às pesquisas em escala industrial motivados pela possível predição do desempenho deste tipo de escoamento. O modelo tridimensional, turbulento e bifásico usado para a predição do escoamento gás-sólido, consiste num conjunto de equações de conservação da massa e momento para cada uma das fases, formuladas seguindo a aproximação Euleriana-Euleriana. As variáveis fluidodinâmicas foram estimadas pela solução do modelo, com o emprego de correlações empíricas da literatura e disponibilizadas pelo código computacional de CFD, para garantir o fechamento do modelo e sua solução numérica. Desde de que a predição da dinâmica do escoamento complexo, em dutos com alto fluxo ascendente de sólidos não é possível por meio somente de equações fundamentais, a maioria dos modelos requerem entradas empíricas, as quais somente são adquiridas com a experimentação. Estes dados foram obtidos dos estudos de PÄRSSINEN e ZHU (2001). A geometria e a malha numérica estrutural do duto vertical foram geradas pelo software (CAD) ICEM, subdividido em DDN (geometria) e Hexa (malha). A adaptação do modelo matemático para a geração do modelo numérico foi alcançada com o uso do simulador comercial CFX 5.7. Os resultados obtidos foram avaliados com respeito à teoria apresentada ao longo da dissertação, sendo finalmente feita uma comparação entre as predições numéricas com o modelo e dados experimentais da literatura
Abstract: The present research presents the mathematical modeling and the simulation of an operation of fast fluidization in a long vertical cylindrical duct line, which is an important operation in many industrial processes, where its main application is tied with the catalytic cracking of oil for gasoline synthesis. It has as objective the attainment of a continuous understanding and knowledge of the development of the gas-solid flow, as well as adding of experiences to research of industrial scale equipments motivated by the possibility of prediction of the performance of this type of flow. The three-dimensional, turbulent and two-phase model used for the prediction of the gas-solid flow, consists of a set of conservation equations of mass and momentum for each phase, which was formulated following the Eulerian-Eulerian approach. The fluid dynamics variables was estimated by the solution of the model with the use of correlations found in the literature and available in the computational CFD code, in order to guarantee the closure of the model and its numerical solution. Since the prediction of the dynamics of the complex flow in ducts with high ascending solid flow is not possible by solely using the basic equations, the majority of the models require the setting empirical, parameters which are acquired only with experimentation. These data were obtained from the studies of PÄRSSINEN and ZHU (2001). The geometry and the structural numerical mesh of the vertical duct was generated by the software (CAD) ICEM, subdivided in DDN (geometry) and Hexa (meshing). The adaptation of the mathematical model for the numerical model generation was reached with the use of the commercial simulator CFX 5.7. The results were evaluated with respect to the theory presented along this dissertation, where comparisons between the numerical predictions with the model and experimental data from the literature were performed.
Mestrado
Desenvolvimento de Processos Químicos
Mestre em Engenharia Química